Station Blackout Analysis of Lungmen ABWR Using TRACE

2012 ◽  
Author(s):  
Jong-Rong Wang ◽  
Hao-Tzu Lin ◽  
Hsiung-Chih Chen ◽  
Wei-Chen Wang ◽  
Chunkuan Shih
Keyword(s):  
Nuclear Power ◽  
Thermal Power ◽  
Electrical Power ◽  
Heat Removal ◽  
Fuel Temperature ◽  
Core Flow ◽  
The Core ◽  
Ac Power ◽  
Station Blackout ◽  
Core Cooling

The Lungmen NPP is the first ABWR (Advanced Boiling Water Reactor) nuclear power plant in Taiwan, consisting of two identical units with 3,926 MWt rated thermal power each and 52.2×106 kg/h rated core flow. The core of Lungmen NPP has 872 bundles of GE14 fuel. There are 10 reactor internal pumps (RIP) in the reactor vessel, providing 111% rated core flow at the nominal operating speed of 151.84 rad/sec. A station blackout (SBO) is defined as the loss of offsite electrical power concurrent with turbine trip and unavailability of the onsite emergency AC power. These result in the loss of core cooling and heat removal systems that rely on the above AC power for their operation. In this research, the TRACE SBO model of Lungmen ABWR has been developed in order for the analysis of SBO transient. The initial condition of SBO transient is 100% rated power/100% rated core flow. The TRACE’s results show that the reactor fuel temperature has been reached 1088.71 K (the zirconium-water reaction may generate) at about 3200 sec. It indicates that the fuels might be damaged after 3200 sec if the RCIC and ACIWA failed to activate in this transient.

scholarly journals Delayed Station Blackout Event and Nuclear Safety

2015 ◽  
Vol 2015 ◽  
pp. 1-9 ◽  
Author(s):  
Andrija Volkanovski ◽  
Andrej Prošek
Keyword(s):  
Time Delay ◽  
Time Delays ◽  
Nuclear Power ◽  
Electrical Power ◽  
Event Tree ◽  
Power Sources ◽  
Ac Power ◽  
Complete Failure ◽  
Station Blackout ◽  
Safety Systems

The loss of off-site power (LOOP) event occurs when all electrical power to the nuclear power plant from the power grid is lost. Complete failure of both off-site and on-site alternating current (AC) power sources is referred to as a station blackout (SBO). Combined LOOP and SBO events are analyzed in this paper. The analysis is done for different time delays between the LOOP and SBO events. Deterministic safety analysis is utilized for the assessment of the plant parameters for different time delays of the SBO event. Obtained plant parameters are used for the assessment of the probabilities of the functional events in the SBO event tree. The results show that the time delay of the SBO after the LOOP leads to a decrease of the core damage frequency (CDF) from the SBO event tree. The reduction of the CDF depends on the time delay of the SBO after the LOOP event. The results show the importance of the safety systems to operate after the plant shutdown when the decay heat is large. Small changes of the basic events importance measures are identified with the introduction of the delay of the SBO event.

The Analysis of AP1000 Beyond Design Basis DEDVI Accident

2013 ◽  
Author(s):  
Sheng Zhu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Heat Removal ◽  
The Core ◽  
Design Basis ◽  
Injection Line ◽  
Design Basis Accident ◽  
Core Cooling

Double ended break of direct vessel injection line (DEDVI) is the most typical small-break lost of coolant accident (LOCA) in AP 1000 nuclear power plant. This study simulated the DEDVI (without actuation of automatic depressurization system 1–3 stage valves, accumulators and passive residual heat removal heat exchanger) beyond design basis accident (BDBA) to validate the safety capability of AP1000 under such conditions. The results show that the core will be uncovered for about 863 seconds and then recovered by water after gravity injection from IRWST into the pressure vessel. The peak cladding temperature (PCT) goes up to 838.08°C, much lower than the limiting value 1204°C. This study confirms that in the DEDVI beyond design basis accident, the passive core cooling system (PXS) can effectually cool the core and preserve it integrate, and ensure the safety of AP 1000 nuclear power plant.

Experimental Study on the Performance of IIST Passive Core Cooling System

2002 ◽  
Author(s):  
Chin-Jang Chang ◽  
Chien-Hsiung Lee ◽  
Wen-Tan Hong ◽  
Lance L. C. Wang
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Heat Removal ◽  
General System ◽  
System Response ◽  
Water Storage Tank ◽  
The Core ◽  
Third Stage ◽  
Core Cooling

The purpose of this study is to conduct the experiments at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility for evaluation of the performance of the passive core cooling system (PCCS) during the cold-leg small break loss-of-coolant accidents (SBLOCAs). Five experiments were performed with (1) three different break sizes, 2%, 0.5%, and 0.2% (approximately corresponding to 1 1/4”, 2”, and 4” breaks for Maanshan nuclear power plant), and (2) 0.2% and 0.5% without actuation of the first-stage and third-stage automatic depressurization valve (ADS-1 and ADS-3) to initiate PCCS for assessing its capacity in accident management. The detailed descriptions of general system response and the interactions of core makeup tanks (CMTs), accumulators (ACCs), automatic depressurization system (ADS), passive residual heat Removal (PRHR), and in-containment refueling water storage tank (IRWST) on the core heat removal are included. The results show: (1) core long term cooling can be maintained for all cases following the PCCS procedures, (2) the core can be covered for the cases of the 0.2% and 0.5% breaks without actuation of ADS-1 and ADS-3.

Fatigue Evaluation for Thermally Stratified DVI Piping

2012 ◽  
Author(s):  
Hwan Ho Lee ◽  
Joon Ho Lee ◽  
Dong Jae Lee ◽  
Seok Hwan Hur ◽  
Il Kwun Nam ◽  
...  
Keyword(s):  
Temperature Distribution ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Heat Removal ◽  
Local Stress ◽  
Piping System ◽  
Hydraulic Analysis ◽  
Fatigue Evaluation ◽  
Core Cooling ◽  
Thermal Hydraulic Analysis

A numerical analysis has been performed to estimate the effect of thermal stratification in the safety injection piping system. The Direct Vessel Injection (DVI) system is used to perform the functions of Emergency Core Cooling and Residual Heat Removal for an APR1400 nuclear power plant (Korea’s Advanced Power Reactor 1400 MW-Class). The thermal stratification is anticipated in the horizontally routed piping between the DVI nozzle of the reactor vessel and the first isolation valve. Non-axisymmetric temperature distribution across the pipe diameter induced by the thermal stratification leads to differential thermal growth of the piping causing the global bending stress and local stress. Thermal hydraulic analysis has been performed to determine the temperature distribution in the DVI piping due to the thermal stratification. Piping stress analysis has also been carried out to evaluate the integrity of the DVI piping using the thermal hydraulic analysis results. This paper provides a methodology for calculating the global bending stresses and local stresses induced by the thermal stratification in the DVI piping and for performing fatigue evaluation based on Subsection NB-3600 of ASME Section III.

scholarly journals Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

2008 ◽  
Vol 2008 ◽  
pp. 1-11 ◽  
Author(s):  
Avinash J. Gaikwad ◽  
P. K. Vijayan ◽  
Sharad Bhartya ◽  
Kannan Iyer ◽  
Rajesh Kumar ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Transport ◽  
Nuclear Power ◽  
Natural Circulation ◽  
Heat Removal ◽  
Transport Systems ◽  
Critical Examination ◽  
Decay Heat ◽  
Station Blackout

Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP) safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs), like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs) is resorted to mitigate consequences of station blackout (SBO). In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR) system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT), SGs, and PDHRs) under station blackout depends on the corresponding system's coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections). On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

The Effect of IRWST Simulation Method on the Analysis of Small Break LOCA Using RELAP5

2017 ◽  
Author(s):  
Caihong Xu ◽  
Guobao Shi ◽  
Kemei Cao ◽  
Xiaoyu Cai ◽  
Zhanfei Qi
Keyword(s):  
Nuclear Power ◽  
Heat Removal ◽  
Simulation Method ◽  
Horizontal Direction ◽  
Water Storage Tank ◽  
Initiation Phase ◽  
Injection Flow ◽  
Loss Of Coolant Accident ◽  
Perfect Mixing ◽  
Core Cooling

The In-containment Refueling Water Storage Tank (IRWST) provides low-pressure safety injection flow for passive CAP1400 Nuclear Power Plant (NPP) during Loss-Of-Coolant-Accident (LOCA) and subsequent Long Term Core Cooling (LTCC). The Passive Residual Heat Removal Heat Exchanger (PRHR HX) and the spargers of Automatic Depressurization System (ADS) stage 1∼3 are submerged in the IRWST. During small break LOCA, heat and mass are delivered through PRHR HX and ADS spargers to IRWST, and IRWST is heated up before its safety injection. However, numerical and experimental investigation has shown that IRWST is not perfect mixing, and thermal stratification exists. During ADS-4/IRWST initiation phase, the temperature of IRWST injection flow is of great importance, and is affected greatly by IRWST simulation method when modeling with system code like RELAP5. In this paper, two different IRWST simulation methods where one use multi channels in horizontal direction while the other use only one, are analyzed for CAP1400 SBLOCA with RE-LAP5, and their effects are compared. Finally, the better method which uses only one channel in horizontal direction is recommended.

Application of LAPUR6 to Lungmen ABWR Stability Analysis

2013 ◽  
Vol 284-287 ◽  
pp. 1146-1150 ◽  
Author(s):  
Hao Tzu Lin ◽  
Jong Rong Wang ◽  
Chun Kuan Shih
Keyword(s):  
Stability Analysis ◽  
Nuclear Power ◽  
Thermal Power ◽  
Core Stability ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Design Features ◽  
Water Reactor ◽  
The Core ◽  
Stability Performance

Lungmen nuclear power plant (NPP) is the first ABWR (Advanced Boiling Water Reactor) in Taiwan and still under construction. It has two identical units with 3,926 MWt rated thermal power each and 52.2×106 kg/hr rated core flow. The core has 872 bundles of GE14 fuel, and the steam flow is 7.637×106 kg/hr at rated power. According to the chapter 4 of Lungmen NPP FSAR (Final Safety Analysis Report), the design features of Lungmen NPP improve the core stability performance and assure that it is more stable than the current BWR (Boiling Water Reactor) NPP in the normal operating regions. In this research, the LAPUR6 stability analysis of Lungmen NPP is performed in order to versify the design features of Lungmen NPP which causes the more stable than the current BWR NPPs. The analysis results of LAPUR6 indicate that the design features of Lungmen NPP can improve the core stability performance effectively and result in the more stable than the current BWR NPPs.

scholarly journals IRRADIATION CHARACTERISTIC OF NATURAL UO2 PIN PHWR TARGET AT PRTF OF RSG – GAS CORE

2017 ◽  
Vol 19 (2) ◽  
pp. 71
Author(s):  
Jati Susilo ◽  
Tagor Malem Sembiring ◽  
Winter Dewayatna
Keyword(s):  
Neutron Flux ◽  
Nuclear Power ◽  
Irradiation Time ◽  
Thermal Power ◽  
Analytical Tool ◽  
Test Facility ◽  
Nuclear Power Reactor ◽  
Ramp Test ◽  
The Core ◽  
Fuel Pin

The RSG-GAS reactor has a facility for irradiation of the fuel pin of nuclear power reactor, namely Power Ramp Test Facility (PRTF). The in-house fabrication PWR fuel pin has prepared for irradiations in the PRTF facility, currently, while the various enrichments of uranium are analyzed using the analytical tool. In the next step, it is planned to perform an irradiation of PHWR fuel pin sample of natural UO2 in the facility. Before irradiation in the core, it should be analyzed by using the analytical tool. The objectives of this paper are to optimize irradiation time based on the burn-up, the generated linear power and the neutron flux level at the target. The 3-dimension calculations have been carried out by using the CITATION code in the SRAC2006 code system. Since the coolant of the reactor is H2O, the effect of moderators in the pressurized tube, H2O and D2O, were analyzed, as well as pellet radius and moderator densities. The calculation results show that the higher linear power as irradiation time longer is occurred preferably in the D2O moderator than in H2O. For the D2O moderator, the higher pressure affects the lower density and longer irradiation time. The maximum irradiation time for natural UO2 fuel pin with the pressurized D2O moderator is about 9.5×104 h, with the linear power of 700 W/cm. During irradiation, neutronic parameters of the core such as excess reactivity and ppf show a very small change, still far below design value.Keywords:  PHWR, Neutron Flux, Thermal Power, PRTF, RSG-GAS KARAKTERISTIK IRADIASI TARGET PIN PHWR UO2 ALAM PADA PRTF TERAS RSG – GAS. Teras RSG-GAS dilengkapi dengan fasilitas untuk uji iradiasi bahan bakar nuklir atau disebut dengan Power Ramp Test Fasility (PRTF). Saat ini sedang dilpersiapkan untuk dilakukan uji sample pin bahan bakar PWR pada fasilitas PRTF. Analisis terhadap uji iradiasi sample pellet UO2 dengan berbagai pengkayaan telah dilakukan menggunakan paket program komputer. Dimasa yang akan datang, uji iradiasi pin bahan bakar PHWR UO2 alam juga sedang dalam perencanaan. Sebelum diiradiasi di dalam teras, maka terlebih dahulu harus dilakukan analisis dengan menggunakan paket program komputer. Tujuan dari penelitian ini adalah optimasi uji iradiasi pin bahan bakar UO2 alam sebagai fungsi waktu iradiasi berdasarkan burn-up, daya linier dan fluks neutron. Perhitungan teras RSG-GAS dilakukan dengan paket program SRAC2006 modul CITATION dalam bentuk geometri 3 dimensi. Analisis dilakukan terhadap pengaruh penggunaan jenis moderator pada tabung tekan iradiasi (H2O dan D2O), perubahan ukuran pelllet UO2 dan perubahan besarnya densitas moderator D2O. Dari analisis hasil perhitungan diketahui bahwa semakin lama waktu iradiasi akan menghasilkan daya termal yang semakin besar jika menggunakan moderator D2O dibandingkan H2O. Semakin tinggi tekanan atau semakin kecil densitas moderator, maka akan menghasilkan daya termal yang semakin besar seiring bertambah lamanya waktu iradiasi. Batas maksimal waktu iradiasi untuk pin bahan bakar UO2 alam dengan moderator D2O bertekanan adalah sekitar 9,5×104 jam, dengan batasan daya linier desain kemampuan peralatan, 700 W/cm. Selama iradiasi, nilai parameter neutronik teras reaktor seperti reaktivitas lebih dan ppf hanya menunjukkan perubahan yang sangat kecil, masih jauh dibawah batas yang ditetapkan dalam desain.Kata kunci: PHWR, Fluks Neutron, Daya Termal, PRTF, RSG-GAS

Maanshan PWR Loss of Flow Transients Analysis With TRACE

2009 ◽  
Author(s):  
Jong-Rong Wang ◽  
Hao-Tzu Lin ◽  
Wei-Chen Wang ◽  
Yi-Hsiang Cheng ◽  
Chunkuan Shih
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Nuclear Power ◽  
Thermal Power ◽  
Complete Loss ◽  
Transient Condition ◽  
Partial Loss ◽  
Core Flow ◽  
System Parameters ◽  
Trace Model

TRACE model of Maanshan Nuclear Power Plant (three-loop PWR) was used to analyze Loss of Flow transient as defined in FSAR Chapter 15. The results were compared with those from RETRAN02 and LOFTRAN/THINC licensing analysis of Westinghouse Inc. Three different initiation events were involved in this analysis: Partial Loss of Flow (PLOF), Complete Loss of Flow-Under Voltage (CLOF-UV) and Complete Loss of Flow-Under Frequency (CLOF-UF). This paper compared important thermal hydraulic parameters at steady state, such as the pressure of pressurizer, cold-leg temperature, and the pressure of steam generator, etc.. It also compared system parameters under transient conditions, such as core thermal power, core flow rate, and pressure of pressurizer, etc.. It is concluded that the steady state results of TRACE calculations are in general good agreements with those from RETRAN02 and have a largest error of 3.03% in the steam generator flow. For transient condition, TRACE results are also comparable with those from LOFTRAN and RETRAN02. In summary, our studies show that Maanshan TRACE model is correct and accurate enough for future safety analysis applications.

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